The present invention relates to plasma devices, particularly devices of the tokamak type used in connection with the study and generation of thermonuclear fusion energy. More particularly, the invention relates to a geometry-insensitive high capacity electric field divertor pump that may used to drive plasma through a desired entrance aperture.
Fusion is the energy source of the sun and other stars. While science has not yet advanced sufficiently far to allow fusion to be used as a practical energy source, scientists and engineers, working at laboratories around the world, are making great strides relative to fusion research and to the engineering development of fusion for electrical power and other applications. Advantageously, fusion fuel is in abundant supply, and the generation of fusion energy provides a safe and clean energy source.
In generating fusion energy, the atoms of two or more fuels, typically deuterium (.sup.2 H) and tritium (.sup.3 H), heavy hydrogen isotopes, are exposed to extremely high temperatures. Such high temperatures separate the positively charged nuclei of the hydrogen isotopes from their normally tightly bound negatively charged electrons, forming a plasma. (A plasma is a hot ionized gas.) When this separation occurs, the neutrons and protons of the nuclei recombine to form a heavier element, such as .sup.4 He, and a neutron or other small nuclear particle. Energy from this reaction is released as kinetic energy of the fast moving reaction products, and it can be converted to heat. The heat thus created provides the high temperature needed to sustain the fusion reaction, and portions thereof can be extracted and used as a useful energy source, e.g., to generate electricity.
The conditions for the fusion reaction are very difficult to achieve. For example, in order to kindle a deuterium-tritium fusion fire, the temperature of the fuel must be heated to over 50,000,000.degree. C. Moreover, to sustain the fusion fire, i.e., to keep the fusion reaction going, it is necessary to confine the normally chaotic mass of fast moving, superheated nuclei (the plasma) long enough for the fuel to react and produce energy beyond that necessary to sustain the temperature. To produce enough fusion reactions to make the process worthwhile, the heat losses from the fuel must be low enough so that the fuel can sustain a temperature of around 150,000,000.degree. C. One such a self-sustaining reaction is achieved, it is possible to use the heat thus produced to generate electricity, or for other purposes.
Achieving such high temperatures requires supplying energy to the fuel and raising its temperature to a level where the internal fusion reactions can provide further heating. Various techniques are currently used to accomplish such heating, e.g., heating with an internal electric current, heating by various waves, and heating by the injection of energetic neutralized hydrogen atoms ("neutral beam injection").
Unlike the sun and stars, where the massive plasma ball is confined by gravity, fusion reactors require some type of container for holding the 50,000,000.degree. C. plasma fireball in a way that prevents it from touching the container walls. (Plasma, which has a density approximately 100,000 times lower than atmospheric pressure, is a mere puff of gas that would quickly cool if it touched the container walls.)
Fortunately, because plasma is an ionized gas, it can be confined with a magnetic field. That is, the otherwise random motion of the charged particles that are found within plasma may be converted to an orderly form of motion that follows the magnetic field lines of an applied magnetic field. Thus, various types of "magnetic bottles" have been developed in the art to create the appropriate magnetic field lines to confine the plasma to a desired volume.
One of the most highly developed magnetic bottles is a toroidal bottle known as the "tokamak". Tokamaks were first developed during the 1960s in the USSR, and have subsequently been adopted as the leading magnetic confinement device. A tokamak includes both external toroidal-field coils and poloidal-field coils that generate magnetic fields, as well as means for generating a toroidal electrical current that flows through the plasma itself. The magnetic fields created by such toroidal- and poloidal-field coil currents, as well as by the plasma electric current, all combine to confine the plasma to a general toroidal shape that encircles a major axis of the tokamak. The poloidal-field coils are also used to magnetically shape the general cross section of the plasma. Tokamaks are well documented in the literature. See, e.g., Artsimovich, L. A., Nuclear Fusion, Vol. 12, pp. 215 et seq. (1972); and Furth, H. P., Nuclear Fusion, Vol. 15, pp. 487 et seq. (1975).
One of the housekeeping tasks associated with the operation of a tokamak is the efficient removal of spent plasma, or plasma exhaust. The external poloidal-field coils of a tokamak may advantageously be used to create one or more "poloidal divertors". See, e.g., Shimomura et al., Physics of Fluids, vol. 19, pp. 1635 et seq. (1976). Broadly stated, a poloidal divertor sets up a magnetic field that manages the plasma exhaust. More specifically, a poloidal divertor guides a thin boundary layer of escaping plasma, known as the scrape-off layer (SOL), located just outside the separatrix, along magnetic lines to solid "divertor targets." (The "separatrix", explained more fully below, is the magnetic surface defined by the applied magnetic fields that separates plasma confined by the magnetic field from plasma not confined by the magnetic field. Figuratively, the separatrix is the "lip" of a magnetic "jar".) Such targets are designed to absorb high heat and particle flux. Unfortunately, it is difficult to design a divertor target that can withstand the spatially peaked thermal loads that occur during a fusion reaction, particularly a long or sustained fusion reaction. Hence, it is preferable that the fusion reactor employ some means for "sweeping" the divertor strike point across the target so as to reduce the heat flux by time averaging.
As the plasma strikes the targets, it cools and neutralizes, becoming a gas, comprised of hydrogen isotopes, helium ash and contaminants eroded from the divertor targets and walls. Such gas needs to be pumped away from the targets in order to prevent its buildup, which buildup (if allowed to continue) would quench the hot plasma. Such buildup has not heretofore been a major problem because the fusion reaction experiments performed to date have been of short duration. However, as the experiments become of longer and longer duration, eventually leading to a continuous nuclear reaction, the need to efficiently remove the neutralized plasma away from the divertor targets will become particularly acute.
Unfortunately, due to the very low pressure associated with the neutral gas near the targets (usually &lt;1 milliTorr), adequate pumping of the gas would require many large ducts through the tritium breeding blanket, radiation shield, and magnet coils of the tokamak. Such ducts would not only take up valuable space in the tokamak, but would also significantly complicate the radiation shielding. What is needed, therefore, is a plasma pump that can efficiently pump the plasma exhaust away from the targets without the need for a large number of massive ducts.
Disadvantageously, pumping of the plasma exhaust (ash and contaminants) is made even more difficult when the desirable high-plasma-confinement operating mode (H-mode) of the tokamak is employed. This is because in the H-mode the plasma retains all ion species for a relatively longer period of time, and the gas pressure near the divertor targets is thus correspondingly less. Hence, what is needed is a means of efficiently removing or pumping plasma exhaust even when the tokamak is operating in the H-mode. (The H-mode tokamak operating regime is fully described in the literature, see, e.g., ASDEX Team, Nuclear Fusion, vol. 29, pp. 1959-2040 (1989).
It is known in the art to build a plenum around the divertor targets designed to collect the plasma exhaust through an aperture opened to the divertor targets. The divertor exhaust gas pressure can be favorably increased by optimizing the entrance aperture geometry for minimum gas backstreaming. However, this makes the pressure very sensitive to small variations of the divertor geometry and plasma conditions. Where the divertor strike point is swept across the target, the entrance aperture geometry is dynamic, and the effectiveness of the plenum at collecting the diverted plasma is severely curtailed. What is needed, therefore, is a means of pumping or forcing the plasma exhaust into the aperture of a collecting plenum that is not significantly geometry sensitive, thereby allowing such plasma pump to be used with a dynamic entrance aperture geometry, such as exists when the divertor strike point is swept across the target.
Some experimental work has been done aimed at applying an electric field E to a plasma flow confined by a magnetic field B, and using the resulting E.times.B drift of the plasma particles to divert the plasma in a desired direction. See, e.g., Strait, E. J., "Poloidal Diverter Experiment With Applied E.times.B/B.sup.2 Drift", Nuclear Fusion, Vol. 21, No. 8, pp. 943-51 (1981); and Strait, et al., "Experimental Demonstration of E.times.B Plasma Divertor", Phys. Fluids, Vol. 21, No. 12, pp. 2342-44 (December 1978). Despite some promising data, however, there still remains a need for a practical application of the principles underlying such experiments to the conditions that prevail in high power tokamak divertors.